Effect of He implantation on the microstructure of zircaloy-4 studied using in situ TEM

M. A. Tunes; R. W. Harrison; G. Greaves; J. A. Hinks; S. E. Donnelly

doi:10.1016/j.jnucmat.2017.06.012

Effect of He implantation on the microstructure of zircaloy-4 studied using in situ TEM

M. A. Tunes, R. W. Harrison, G. Greaves, J. A. Hinks, S. E. Donnelly

Research output: Contribution to journal › Article › peer-review

Abstract

Zirconium alloys are of great importance to the nuclear industry as they have been widely used as cladding materials in light-water reactors since the 1960s. This work examines the behaviour of these alloys under He ion implantation for the purposes of developing understanding of the fundamental processes behind their response to irradiation. Characterization of zircaloy-4 samples using TEM with in situ 6 keV He irradiation up to a fluence of 2.7×1017ions·cm−2in the temperature range of 298 to 1148 K has been performed. Ordered arrays of He bubbles were observed at 473 and 1148 K at a fluence of 1.7×1017ions·cm−2in αZr, the hexagonal compact (HCP) and in βZr, the body centred cubic (BCC) phases, respectively. In addition, the dissolution behaviour of cubic Zr hydrides under He irradiation has been investigated.

Original language	English
Pages (from-to)	230-238
Number of pages	9
Journal	Journal of Nuclear Materials
Volume	493
Early online date	13 Jun 2017
DOIs	https://doi.org/10.1016/j.jnucmat.2017.06.012
Publication status	Published - 1 Sept 2017

Keywords

Bubble lattices
Hydrides
In situ ion irradiation
Radiation damage
Zirconium alloys

Access to Document

10.1016/j.jnucmat.2017.06.012

Cite this

@article{d0de9ddea6ee4e0780a93c73c29a50b8,

title = "Effect of He implantation on the microstructure of zircaloy-4 studied using in situ TEM",

abstract = "Zirconium alloys are of great importance to the nuclear industry as they have been widely used as cladding materials in light-water reactors since the 1960s. This work examines the behaviour of these alloys under He ion implantation for the purposes of developing understanding of the fundamental processes behind their response to irradiation. Characterization of zircaloy-4 samples using TEM with in situ 6 keV He irradiation up to a fluence of 2.7×1017ions·cm−2in the temperature range of 298 to 1148 K has been performed. Ordered arrays of He bubbles were observed at 473 and 1148 K at a fluence of 1.7×1017ions·cm−2in αZr, the hexagonal compact (HCP) and in βZr, the body centred cubic (BCC) phases, respectively. In addition, the dissolution behaviour of cubic Zr hydrides under He irradiation has been investigated.",

keywords = "Bubble lattices, Hydrides, In situ ion irradiation, Radiation damage, Zirconium alloys",

author = "Tunes, {M. A.} and Harrison, {R. W.} and G. Greaves and Hinks, {J. A.} and Donnelly, {S. E.}",

year = "2017",

month = sep,

day = "1",

doi = "10.1016/j.jnucmat.2017.06.012",

language = "English",

volume = "493",

pages = "230--238",

journal = "Journal of Nuclear Materials",

issn = "0022-3115",

publisher = "Elsevier BV",

}

TY - JOUR

T1 - Effect of He implantation on the microstructure of zircaloy-4 studied using in situ TEM

AU - Tunes, M. A.

AU - Harrison, R. W.

AU - Greaves, G.

AU - Hinks, J. A.

AU - Donnelly, S. E.

PY - 2017/9/1

Y1 - 2017/9/1

N2 - Zirconium alloys are of great importance to the nuclear industry as they have been widely used as cladding materials in light-water reactors since the 1960s. This work examines the behaviour of these alloys under He ion implantation for the purposes of developing understanding of the fundamental processes behind their response to irradiation. Characterization of zircaloy-4 samples using TEM with in situ 6 keV He irradiation up to a fluence of 2.7×1017ions·cm−2in the temperature range of 298 to 1148 K has been performed. Ordered arrays of He bubbles were observed at 473 and 1148 K at a fluence of 1.7×1017ions·cm−2in αZr, the hexagonal compact (HCP) and in βZr, the body centred cubic (BCC) phases, respectively. In addition, the dissolution behaviour of cubic Zr hydrides under He irradiation has been investigated.

AB - Zirconium alloys are of great importance to the nuclear industry as they have been widely used as cladding materials in light-water reactors since the 1960s. This work examines the behaviour of these alloys under He ion implantation for the purposes of developing understanding of the fundamental processes behind their response to irradiation. Characterization of zircaloy-4 samples using TEM with in situ 6 keV He irradiation up to a fluence of 2.7×1017ions·cm−2in the temperature range of 298 to 1148 K has been performed. Ordered arrays of He bubbles were observed at 473 and 1148 K at a fluence of 1.7×1017ions·cm−2in αZr, the hexagonal compact (HCP) and in βZr, the body centred cubic (BCC) phases, respectively. In addition, the dissolution behaviour of cubic Zr hydrides under He irradiation has been investigated.

KW - Bubble lattices

KW - Hydrides

KW - In situ ion irradiation

KW - Radiation damage

KW - Zirconium alloys

U2 - 10.1016/j.jnucmat.2017.06.012

DO - 10.1016/j.jnucmat.2017.06.012

M3 - Article

SN - 0022-3115

VL - 493

SP - 230

EP - 238

JO - Journal of Nuclear Materials

JF - Journal of Nuclear Materials

ER -

Effect of He implantation on the microstructure of zircaloy-4 studied using in situ TEM

Abstract

Keywords

Access to Document

Fingerprint

Cite this